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  • Articles  (8)
  • tokamak  (8)
  • Springer  (8)
  • American Chemical Society (ACS)
  • 2020-2024
  • 1980-1984  (8)
  • Energy, Environment Protection, Nuclear Power Engineering  (8)
  • Nature of Science, Research, Systems of Higher Education, Museum Science
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  • Articles  (8)
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  • Springer  (8)
  • American Chemical Society (ACS)
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  • Energy, Environment Protection, Nuclear Power Engineering  (8)
  • Nature of Science, Research, Systems of Higher Education, Museum Science
  • Physics  (1)
  • 1
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 15-48 
    ISSN: 1572-9591
    Keywords: Plasma fueling ; pellet injection ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract The injection of frozen pellets composed of the isotopes of hydrogen has become the leading candidate for refueling fusion power reactors based on the tokamak concept. This lofty position has been reached partly as a result of efforts to find an attractive solution to the perplexing problem of depositing atoms of fuel deep within the magnetically confined, hot plasma, and because of some recent experimental successes. To some extent, the relative merits of this technique will depend upon the distance that the cryogenic pellet will penetrate such a plasma, and the early exploratory research has addressed this problem on both theoretical and experimental fronts. The conclusion from the theoretical effort is that a protective blanket consisting of hydrogenic gas or cold plasma will envelope the pellet and partially shield the surface from the intense plasma heat flux. The blanket prolongs pellet lifetime, but penetration to the plasma center might require pellet injection velocities in excess of 10 km/s. The need for central penetration has not yet been established either theoretically or experimentally. The experiments performed to date have verified the existence of a shielding mechanism in general, and pellet ablation models that incorporate neutral gas shielding in particular are in adequate agreement with the experiments. Magnetic shielding effects are expected to contribute to, but not dominate, self-shielding in the higher plasma temperature regimes of the future. The tokamak plasma has demonstrated a surprising resilience even to massive density perturbations caused by the large refueling pellets used in present experiments. The characteristic discharge behavior is qualitatively not unlike that observed with gas puffing; but, for the first time, central plasma fueling has been studied, and this does not appear to be superior to refueling by partial pellet penetration. If relatively large pellets containing a significant fraction of the total plasma charge are acceptable in the present resistive plasma regimes, then it can be argued that they should have little impact on the gross stability of a hot thermonuclear tokamak plasma. Large pellets are preferable from the standpoint of attaining deep penetration, and this has important implications for the technology of pellet injection. The interesting velocity regime of 1 km/s has already been achieved with simple gun-type devices and this should be adequate for near-term tokamak experiments. Further improvements are anticipated, but the 10 km/s and above regime is uncertain; and, if current theory and experiments extrapolate to the future, such velocities might be desirable but unnecessary.
    Type of Medium: Electronic Resource
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  • 2
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 1 (1981), S. 111-116 
    ISSN: 1572-9591
    Keywords: Ignition ; copper magnets ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Design considerations have been developed for a compact ignition test reactor (CITR). The objectives of this tokamak device are to achieve ignition, to study the characteristics of plasmas that are self-heated by alpha particles, and to investigate burn control. To achieve a compact design, the toroidal field magnet consists of copper-stainless steel plates to accommodate relatively high stresses; it is inertially cooled by liquid nitrogen. No neutron shielding is provided between the plasma and the toroidal field magnet. The flat-top of the toroidal field magnet is ∼10 s. Strong auxiliary heating is employed. In one design option, adiabatic compression in major radius is employed to reduce the neutral beam energy required for adequate penetration; thiscompression boosted design option has a horizontally elongated vacuum chamber; illustrative parameters are a compressed plasma witha=0.50 m, R=1.35 m,B T =9.1 T, and a neutral beam power of ∼15 MW of 160 keVD 0 beams. A design option has also been developed for alarge bore device, which utilizes a circular vacuum chamber. Thelarge bore design provides increased margin and flexibility; both direct heating with RF or neutral beam injection and compression boosted startup are possible. The large bore design also facilitates the investigation of high-Q driven operation. Illustrative plasma parameters for full use of the large bore area=0.85 m,R=1.90 m, andB T =7.5 T.
    Type of Medium: Electronic Resource
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  • 3
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 3 (1983), S. 63-66 
    ISSN: 1572-9591
    Keywords: tokamak ; fusion reactor ; pulsed
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract We discuss the design features of a commercial tokamak reactor in which day long pulses are provided by the volt-second capability of the ohmic heating transformer. Illustrative parameters are a major radius of 9.7 m, a minor radius of 1.95 m, an average toroidal beta of 0.036, a magnetic field on axis of 6.1 T, a neutron wall loading of 2.3 MW/m2 and a thermal power level of 4000 MW. The tokamak is modularized into units which consist of two toroidal field coils, blanket and shield and first wall. The removal of the two toroidal field coils associated with a module would be carried out without warming up the rest of the toroidal field coil set.
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  • 4
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 3 (1983), S. 217-229 
    ISSN: 1572-9591
    Keywords: tokamak ; fusion reactors ; tokamak scaling ; ignition
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R 〈 1m). The relevant machine technology is discussed.
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  • 5
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 2 (1982), S. 197-205 
    ISSN: 1572-9591
    Keywords: fusion reactors ; tritium breeding ; semicatalyzed deuterium ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Nontritium-breeding D-T reactors have decisive advantages in minimum size, unit cost, variety of applications, and ease of heat removal over reactors using any other fusion cycle, and significant advantages in environmental and safety characteristics over breeding D-T reactors. Considerations of relative energy production demonstrate that the most favorable source of tritium for a widely deployed system of nontritium-breeding D-T reactors is the very large (∼10 GW thermal) semicatalyzed-deuterium (SCD), or sub-SCD reactor, where none of the escaping3He (〉 95%) or tritium (〈 25%) is reinjected for burn-up. Feasibility of the ignited SCD tokamak reactor requires spatially averaged betas of 15 to 20% with a magnetic field at the TF coils of 12–13 T.
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  • 6
    ISSN: 1572-9591
    Keywords: diagnostics ; alpha particles ; tokamak ; fusion diagnostics
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract Methods are proposed for measuring the alpha-particle distribution in magnetically confined fusion plasmas using neutral-atom doping beams, ultraviolet spectroscopy, and neutral particle detectors. In the first method, single charge exchange reactions, A0+He2+→A+ +(He+)*, are used to populate then=2 andn=3 levels of He+. The ultraviolet photons from the decaying excited states are Doppler shifted by 5–10 Å from those produced by the thermalized alpha-particle “ash.” In the second method, double charge exchange reactions, A0+He2+→A2++He0, enable fast neutralized alpha particles to escape from the plasma and be detected by neutral particle analyzers. These methods are distinguished from similar techniques of observing plasma impurities in that, in principle, they allow a determination of the dependence of the distribution function on energy and pitch angle, as well as on spatial position. Detector configurations are analyzed, count rates are estimated, and the detector feasibility is discussed. A preliminary analysis of the feasibility of the required neutral beams is presented, and exploratory experiments on existing devices are suggested.
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  • 7
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 2 (1982), S. 225-236 
    ISSN: 1572-9591
    Keywords: neutral beam deposition ; plasma heating ; tokamak
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A “parametric” pencil beam model is introduced for describing the attenuation of an energetic neutral beam moving through a tokamak plasma. The nonnegligible effects of a finite beam cross-section and noncircular shifted plasma cross-sections are accounted for in a simple way by using a smoothing algorithm dependent linearly on beam radius and by including information on the plasma flux surface geometry explicitly. The model is bench-marked against more complete and more time-consuming two-dimensional Monte Carlo calculations for the case of a large D-shaped tokamak plasma with minor radiusa=120 cm and elongationb/a=1.6. Deposition profiles are compared for deuterium beam energies of 120–150 keV, central plasma densities of 8×1013 to 2×1014 cm−3, and beam orientation ranging from perpendicular to tangential to the inside wall.
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  • 8
    Electronic Resource
    Electronic Resource
    Springer
    Journal of fusion energy 3 (1983), S. 13-24 
    ISSN: 1572-9591
    Keywords: fusion reactors ; current drive ; neutral beams ; tokamak ; internal transformer
    Source: Springer Online Journal Archives 1860-2000
    Topics: Energy, Environment Protection, Nuclear Power Engineering
    Notes: Abstract A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia time scales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.
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